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DOE OSTI2025

Evaluation of New Additions to OLI Software in Predicting Mercuric and Mercurous Species in Liquid Waste Operations

Speciation of mercury during the pretreatment steps of tank waste processing is critical to successful mercury removal prior to vitrification during Liquid Waste Operations (LWO) at SRS. OLI software has been used to predict mercury speciation and activity throughout LWO. The OLI software operates based on a thermodynamic framework called the Mixed Solvent Electrolyte (MSE) framework. The MSE framework allows prediction in theoretically infinitely dilute to concentrated mixtures (e.g., purely solute solutions). Before modification to the MSE framework databanks, certain critical mercury species were missing in the MSE databank, and some thermodynamic data needed to be updated for the OLI software to accurately predict mercury chemical species in SRS waste tanks. To better reflect streams across LWO, new mercury species were integrated into the MSE database. To evaluate the changes to the OLI MSE framework per the Technical Task Request (TTR) and the Task Technical and Quality Assurance Plan (TTQAP), waste stream compositions from Tanks 38, 43, and Tank 50 decontaminated salt solution (DSS) were used as model inputs. Models were developed and executed using both the old and new databases. Compositional analyses from caustic Tank 50 DSS and caustic Tanks 38 and 43 were used as the input streams. These streams represent the most comprehensive chemical data sets where both mercury and tank constituents were measured together. Results for Tank 50 DSS predict HgO as the predominant species in both databases. Both methyl and dimethyl Hg species are present when the new database is ‘on’ and are not predicted with the new database turned ‘off’. The new database predicts a greater amount of HgO and a greater fraction of it in the solid phase. Pourbaix diagrams (potential vs. pH) generated for each Tank 50 DSS were identical regardless of which database was used. Elemental Hg and HgO were predicted in the water stable region under basic conditions. Tanks 38 and 43 follow similar trends as the Tank 50 DSS models. Unlike Tanks 38 and 50 DSS, the Tank 43 Pourbaix plot shows a region of stability for an aqueous HgOHCO3 - species between approximately pH 7-11. In all streams, when MeHg+ is included in the inputs, the new database predicts aqueous MeHgOH as the dominant species. If elemental or dimethyl mercury is in the waste stream, the new database model predicts they are unchanged and remain in those states and quantities. Additionally, the total mercury values are reported for both the measured input data and the OLI output data for all considered tanks. The summary indicates that the percentage error between the measured and calculated values is less than 1% in all cases The reconciliations and generation of the Pourbaix diagrams for Tank 50 DSS took approximately ten times longer with the new database ‘on’. In addition, over the course of that time, models with the new database ‘on’ were more likely to crash or display an error. Some modest performance improvements were noted when modeling with an i7 processor versus an i5. An example error is found in Appendix A. Furthermore, Appendix B provides V&V for two chemical systems analyzed with the OLI software, results were satisfactory. It is recommended to utilize the new databases (i.e., HCO.ddb and SR-Hg.ddb) in future Savannah River Mission Completion applications of OLI to represent pseudo steady-state. Furthermore, the integration and utilization of the new databases (i.e., HCO.ddb and SR-Hg.ddb) in modeling applications (e.g., Aspen) is also recommended.

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DOE OSTI2019

Salt Heater Test (FY19), Rev. 2

This report summarizes the 2019 fiscal year (FY19) status of the borehole heater test in salt funded by the US Department of Energy Office of Nuclear Energy (DOE-NE) Spent Fuel and Waste Science & Technology (SFWST) campaign. This report satisfies SFWST level-three milestone report M3SF-19SN010303033. This report is an update of the April 2019 level-two milestone report M2SF-19SNO10303031 to reflect the nearly complete as-built status of the borehole heater test. This report discusses the fiscal year 2019 (FY19) design, implementation, and preliminary data interpretation plan for a set of borehole heater tests call the brine availability tests in salt (BATS), which is funded by the DOE Office of Nuclear Energy (DOE-NE) at the Waste Isolation Pilot Plant (WIPP), a DOE Office of Environmental Management (DOE-EM) site. The organization of BATS is outlined in Project Plan: Salt In-Situ Heater Test (SNL, 2018). An early design of the field test is laid out in Kuhlman et al. (2017), including extensive references to previous field tests, which illustrates aspects of the present test. The previous test plan by Stauffer et al. (2015) places BATS in the context of a multi-year testing strategy, which involves tests of multiple scales and processes, eventually culminating in a drift-scale disposal demonstration. This level-3 milestone report is an update of a level-2 milestone report from April 2019 by the same name. The update adds as-built details of the heater test, which at the time of writing (August 2019) is near complete implementation.

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DOE OSTI2018

Cementitious Waste Form Development for Treatment of EMF Evaporator Bottoms - 18611

Direct Feed Low-Activity Waste operations at the Hanford Site (Washington State, USA) involve concentrating the Hanford low-activity waste (LAW) melter off-gas condensate waste stream by evaporation at the Effluent Management Facility (EMF) followed by recycle back to the melter feed. The EMF evaporator bottoms waste stream is expected to contain variable, high levels of halides and sulfate that could strain LAW melter performance and co-mingled radionuclides, technetium-99 ({sup 99}Tc) and iodine-129 ({sup 129}I), that readily volatilize in the LAW melter and are expected to accumulate during recycle. To this end, there is a need for alternative disposition paths for the EMF evaporator bottoms waste stream that bypass recycling to the LAW melter. One such alternative is solidification in low-temperature cementitious waste forms (CWFs) that meet disposal facility acceptance criteria. The work herein addresses recent waste form development and testing performed to identify CWF candidates that meet off-site disposal acceptance criteria and/or on-site Hanford Integrated Disposal Facility waste acceptance criteria. Phase one testing activities included the production and evaporation of simulated EMF evaporator feed from the Hanford Waste Treatment and Immobilization Plant melter offgas effluent liquid secondary waste stream, which was then analyzed to demonstrate the behavior and fate of contaminants of concern (COCs). Based on analyses from these preliminary tests, a simulant for the EMF evaporator bottoms waste stream was formulated and used to evaluate several dry material compositions proposed for CWF development and testing. In particular, dry ingredient recipes used in early testing efforts contained a combination of ordinary Portland cement (OPC), Class F fly ash (FA), blast furnace slag (BFS), Aquaset{sup R} II-H, and/or Aquaset{sup R} II-GH to generate CWFs at various water-to-dry material ratios. CWFs were then tested according to the Toxicity Characteristic Leaching Procedure (TCLP) for COCs, such as Cr, Zn, As, and Se, to determine which specimens comply with land disposal restrictions (40 CFR 268). The results of TCLP testing showed that chromium leaching was a persistent challenge, except when using the original Cast Stone dry ingredient recipe (8% OPC, 47% BFS, 45% FA) to immobilize the EMF evaporator bottoms simulant. Improving Cr immobilization and evaluating leaching behavior for COC Hg were identified as objectives for the second phase of testing. During second phase testing efforts, a screening test was performed to determine whether the bulk EMF waste composition is a factor in producing an acceptable solidified waste form. A series of eight simulants with varying concentrations of major salts of boron, chloride, nitrite, and sulfate were prepared and immobilized using three dry ingredient recipes. Each dry ingredient recipe contained two to three of the following: BFS, OPC, FA, and Aquaset{sup R} II-GH. All eight simulants were spiked with Zn, Cr, As, Se, and Hg prior to production of the CWFs, similar to previous tests, and used to test for residual free liquids and TCLP analysis. The TCLP results indicate that all phase two formulated CWFs meet land disposal requirements when compared to Universal Treatment Standards (40 CFR 268) for hazardous wastes, regardless of simulant composition. Furthermore, re-absorption of residual free liquids is most quickly achieved for nearly all simulant compositions using the original Cast Stone recipe; however, replicas analysis is required to corroborate these findings. (authors)

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DOE OSTI2026

Application of Ibuprofen Sodium Dihydrate for Thermochemical Energy Storage

Thermochemical energy storage (TCES) offers a transformative approach to address grid instability by harnessing reversible chemical reactions for efficient heat storage and release. Here, we introduce pharmaceutical organic salt hydrates as a class of materials with exceptional performance for low-grade waste heat recovery. We demonstrate ibuprofen sodium dihydrate (ISD) as an example organic hydrate exhibiting a dehydration temperature range of 60-110 °C and a remarkable dehydration enthalpy of up to 59.5 kJ/mol of water, ideally suited for capturing industrial and residential waste heat. Using rigorous multimodal characterization, including thermogravimetric analysis, differential scanning calorimetry, in-situ FTIR, in-situ PXRD, and NMR, we demonstrate ISD's superior thermal, chemical, and structural stability over 150 hydration-dehydration cycles, achieving an unprecedented cycling efficiency of ~99.9%. Compared to conventional inorganic salt hydrates like strontium chloride hexahydrate and calcium oxalate monohydrate, ISD showcases enhanced durability without deliquescence or pulverization, even under high-humidity conditions. In-situ analyses confirm the transition from ISD to ibuprofen sodium anhydrous (ISA) proceeds with structural reorganization, thereby combining the dehydration mechanism with phase transitions, resulting in higher energy storage capacity. Microstructural analyses reveal that repeated water intercalation and structural transitions aid in creating significant porosity that enhances water transport kinetics, further improving the hydration/dehydration performance. By combining the phase change and chemical dehydration mechanisms, ISD paves the way for designing a new class of organic salt hydrates, offering tunable properties to meet diverse thermal energy storage demands and supporting sustainable grid resilience.

Thangaraj, Kavin C.↗

DOE OSTI2018

Performance Assessment of a Generic Nuclear Waste Repository in Shale - 18392

A performance assessment (PA) is an important component of a comprehensive safety analysis for a nuclear waste repository. In a PA, probabilistic simulations of the total repository system are performed, and results are evaluated against performance metrics. Uncertainty and sensitivity analyses help prioritize additional research and model development. The United States Department of Energy (DOE) has been developing a state of the art PA simulation software tool-kit, Geologic Disposal Safety Assessment (GDSA) Framework, that couples increasingly higher fidelity models of subsystem processes into total system PA simulations. Over the past several years, PAs of generic repositories in three geologic media (salt, shale, and crystalline rock) have demonstrated ongoing developments in capability. The current PA of a nuclear waste repository in a generic shale formation showcases GDSA Framework, including capabilities in domain discretization (Cubit), multi-physics simulations (PFLOTRAN), uncertainty and sensitivity analysis (Dakota), and visualization (Paraview). The generic shale reference case considers the disposal of 22,000 metric tons heavy metal of commercial spent nuclear fuel (SNF) in a generic shale formation. PA simulations account for the thermal load and radionuclide inventory of the waste form, components of the engineered barrier system including waste package and buffer, and components of the natural barrier system including the host rock shale and underlying and overlying stratigraphic units. Two repository layouts are considered, one for emplacement of waste packages containing 12 pressurized water reactor (PWR) assemblies, and one for 4-PWR waste package emplacement. Model domains are half-symmetry and contain between 7 and 22 million grid cells. Grid refinement captures the detail of individual waste packages, emplacement drifts, access drifts, and shafts. Simulations are run in a high-performance computing (HPC) environment on 512 or 2048 processes. The governing equations describing coupled heat and fluid flow and reactive transport are solved with PFLOTRAN, an open-source, massively parallel multiphase flow and reactive transport code. Additional simulated processes include waste package degradation; waste form dissolution; radioactive decay and ingrowth in aqueous, solid, and adsorbed phases; and a simple biosphere based on a pumping well. Simulations are run to 106 y; dose and radionuclide concentrations are observed within aquifers at a point approximately 5 km downgradient of the repository. Dakota is used to sample likely ranges of input parameters including waste form and waste package degradation rates and properties of engineered and natural materials to quantify uncertainty in predicted concentrations and sensitivity to input parameters. Given the assumptions of the reference case, results of 12-PWR and 4-PWR simulations are very similar. Dose at the pumping well 5-km downgradient of the repository does not exceed 10-9 Sv/y. I-129 is the largest contributor to dose. The dose due to Cl-36, the next largest contributor, is orders of magnitude smaller. At the pumping well, the concentration of I-129 5-km downgradient does not exceed 10-14 mol/L. Rank correlation coefficients indicate that the maximum concentration of I-129 within the aquifer is sensitive to shale (repository host rock) porosity and aquifer permeability. These results affirm that HPC-capable codes can be used to simulate important multi-physics couplings directly in a total system performance assessment of a geologic repository, and can be used in prioritization of future research and model development. (authors)

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DOE OSTI2018

LANL Salt International Update

This milestone presents material relating to international salt investigations during the 2018 fiscal year. Overall salt efforts this year were divided between the currently in-progress borehole experiments (see test plan TP-18-001, rev. 0 and DOE Milestone M3SF-18LA010303014), model development (see DOE milestone M3SF-18LA010303015), and validation of numerical modeling for generic salt repository applications domestically (intranational) and abroad (international). The present milestone focuses on the last of these subject areas. Work for international dissemination was conducted during the first half of the 2018 fiscal year and includes two international conference presentations - Geological Society of America (GSA) and the International Society for Porous Media (IntePore) - and a manuscript which is currently in review in Vadose Zone Journal.

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DOE OSTI2023

Recovery of Enriched Mo-100 from Low Specific Activity Mo-99/Tc-99m Generator Waste Streams

This study assessed two potential techniques for the extraction of pure Mo-100 target material from Mo-99/Tc-99m waste streams: 1) an established solvent-extraction method tailored for Mo-99 spent solution recycling, and 2) a low-temperature molten salt electrochemical process. By making minor modifications to the conventional solvent extraction method, particularly to accommodate the altered solubility limits arising from elevated Na+ concentrations in the waste, we achieved quantitative Mo recoveries. Although the electrochemical technique offers advantages in scalability and processing simplicity, it proved ineffective in recovering Mo from a NaOH-KOH salt melt.

07 ISOTOPE AND RADIATION SOURCES↗

DOE OSTI2023

Ab-Initio molecular dynamics simulations of binary NaCl-ThCl 4 and ternary NaCl-ThCl 4 -UCl 3 molten salts

Molten salt reactor (MSR) with Thorium (Th) fuel cycle has attracted growing attention due to its merits such as safety, low radioactive waste production, and non-proliferation. The design and safe operation of MSR rely on a thorough understanding of the thermophysical properties of molten salts over a wide range of composition and temperature. However, experimental data of Th-based molten salts are limited due to the inherent challenges of dealing with corrosive molten salts and the radioactive nature of actinides. Here, in this work, thermophysical properties including density, heat capacity, thermal expansion coefficient, and mixing energy of binary NaCl-ThCl 4 and ternary NaCl-ThCl 4 -UCl 3 molten salts are explored using ab-initio molecular dynamic simulations (AIMD) with the dDsC dispersion correction. The calculated mixing energy of binary NaCl-ThCl 4 exhibits a minimum close to the eutectic composition. The heat capacity of the mixtures is linearly dependent on the mole fraction of each component. 7-fold and 6-fold coordinated Th complexes are dominant in the mixtures. 8-fold coordinated Th complex increases with ThCl 4 fraction due to the formation of network structures. The average coordination number of Th exhibits a minimum near the eutectic composition. The minimum mixing energy for the ternary mixtures is observed in systems with a composition close to [NaCl] 0.5 [ThCl 4 ] 0.25 [UCl 3 ] 0.25 . The density positively deviates from the ideal solution in mixtures near this composition. These results are important to fill the data and knowledge gap of ThCl 4 molten salts.

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DOE OSTI2018

Innovative Plutonium Waste Forms: Acceptability to Shipp to WIPP

The mission of the Waste Isolation Pilot Plant in New Mexico is the safe and permanent disposition of transuranic radioactive waste from atomic energy defense activities of the United States. The plant consists of an underground repository in a salt formation that is free of flowing water, easily mined, impermeable, and geologically stable. The salt rock naturally seals fractures and closes openings over time. The repository is now anticipated to receive a greater amount of plutonium materials than originally envisioned. An engineered waste form for safe and efficient transport and long-term disposal of plutonium materials is an urgent need. An optimal waste form could also reduce the disposition life cycle and its associated costs to the taxpayer.

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DOE OSTI2022

Stable salt hydrate-based thermal energy storage materials

Heating and cooling systems in building infrastructure utilize conventional materials that account for a considerable amount of energy usage and waste. Phase change material (PCM) is considered a promising candidate for thermal energy storage that can improve energy efficiency in building systems. In this work, a novel salt hydrate-based PCM composite with high energy storage capacity, relatively higher thermal conductivity, and excellent thermal cycling stability was designed and developed. The thermal cycling stability of the PCM composite was enhanced by using dextran sulfate sodium (DSS) salt as a polyelectrolyte additive, which significantly reduced the phase segregation of salt hydrate. The energy storage capacity and the thermal conductivity of the composite were enhanced by the addition of various graphitic materials along with Borax nucleator. A significant increase in thermal cycling stability was observed for the DSS-modified composite, with over 100 thermal cycles without degradation. The final PCM composite exhibited as much as 290% increase in energy storage capacity relative to the pure salt hydrate, and approximately 20% increase in thermal conductivity. In addition, the PCM composite developed can be produced at larger scale, and can potentially change the future of heating/cooling system in building infrastructure.

42 ENGINEERING↗

DOE OSTI2018

Mass transport analysis for tritium removal in FHRs

Tritium control and mitigation is one of the most significant issues in Fluoride Salt-cooled Hightemperature Reactors (FHRs). To address this issue, a cross-flow tritium removal facility has been proposed to remove molecular tritium, T 2 , from the primary coolant. To model the performance of the design, a method is developed based on the logarithmic mean difference of the square root of the tritium partial pressures on the upstream side and the purging gas side. A computer code has been developed based on this global analysis method to study the effectiveness of tritium removal in such a cross-flow tritium removal facility. The code is benchmarked against experimental data from a hydrogen separation experiment in the literature and compared with the results from a code based on traditional finite volume method. Evaluation of the cross-flow tritium removal facility designed for FHR systems has also been performed. The results show how the key factors, i.e., the fluid flow rates and tritium inlet concentration affect the performance of the facility.

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DOE OSTI2020

Characterization of Tank 11H Samples from Tank Closure Cesium Removal (TCCR) Batch 3 - Intermediate and Final Samples

11H↗

DOE OSTI2019

Solid-liquid Separation Testing for the Remediation of Hanford Waste Treatment Plant Low Activity Waste Melter Off-Gas Condensate

The Low Activity Waste (LAW) vitrification facility at the Hanford Waste Treatment and Immobilization Plant (WTP) will generate an aqueous condensate stream (LAW Off-Gas Condensate) from the off-gas system. The plan for disposition of this stream during Direct Feed LAW operations is to transfer it to the Effluent Management Facility (EMF), where it will be concentrated by evaporation and recycled to the LAW vitrification facility again. The primary reason to recycle this stream is so that the semi-volatile 99Tc isotope eventually becomes incorporated into the glass. However, this stream also contains non-radioactive salt components that are problematic in the melter, so diversion of this stream to another process would eliminate recycling of these salts and would enable simplified operation of the LAW melter and the Pretreatment Facilities. This diversion from recycling this stream within WTP would have the effect of decreasing the quantity of glass waste produced by allowing further increases in waste loading, resulting in a shortened LAW vitrification mission duration. The concept being tested here involves selectively removing the 99 Tc so that the decontaminated aqueous stream, with the problematic salts, can be disposed as a different waste form and avoid reprocessing through vitrification.

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DOE OSTI2024

Experimental Investigations into the Corrosion of Alloy 625 Using NaCl-PuCl3 Molten Salt in a Natural Circulation Microloop

Molten salt reactors (MSRs) can potentially revolutionize the nuclear industry by providing a path to a near-zero nuclear waste fuel cycle, contributing to more sustainable energy sources. As a plethora of MSR developers in the United States work toward an aggressive commercialization timeline, many of their fueled-salts—notably, chloride-based compositions—have limited operational testing with nuclear material. Licensing and operating these reactors require an understanding of corrosion effects on reactor materials of construction under operational conditions. The TerraPower Molten Chloride Fast Reactor (MCFR) is a liquid-fueled chloride-salt fast reactor which has received notable interest from the utility sector based on its desirable economic characteristics. The reactor operates at low pressure but does not require the use of highly reactive chemicals, leading to a reduced use of concrete and steel during construction. Additionally, liquid fuel allows for inherently stable behavior and natural circulation during a loss-of-site-power scenario. MCFR can be refueled while operating which makes it compatible with variable generation sources such as wind and solar. MCFR is a breed-and-burn in-situ reactor that does not implement any chemical processing or separations in the fuel cycle. Only mechanical filtration of noble metals and off-gassing of noble gases are utilized while the actinides stay mixed with the fuel at all times. The MCFR will require technology development to reach commercialization. With a breed-and-burn in-situ reactor like MCFR, the transmutation of fertile U-238 to fissile Pu-239 allows for much greater fuel utilization.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2017

Investigation of thermolytic hydrogen generation rate of tank farm simulated and actual waste

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DOE OSTI2019

Investigation of thermolytic hydrogen generation rate of tank farm simulated and actual waste

To support resolution of Potential Inadequacies in the Safety Analysis for the Savannah River Site (SRS) Tank Farm, Savannah River National Laboratory conducted research to determine the thermolytic hydrogen generation rate (HGR) with simulated and actual waste. Gas chromatography methods were developed and used with air-purged flow systems to quantify hydrogen generation from heated simulated and actual waste at rates applicable to the Tank Farm Documented Safety Analysis (DSA). Initial simulant tests with a simple salt solution plus sodium glycolate demonstrated the behavior of the test apparatus by replicating known HGR kinetics. Additional simulant tests with the simple salt solution excluding organics apart from contaminants provided measurement of the detection and quantification limits for the apparatus with respect to hydrogen generation. Testing included a measurement of HGR on actual SRS tank waste from Tank 38. A final series of measurements examined HGR for a simulant with the most common SRS Tank Farm organics at temperatures up to 140 °C. The following conclusions result from this testing. HGR was measured for an actual-waste Tank 38 sample. The HGR measurements at 80, 90, 100, and 110 °C were 1.7×10 -7 , 4.7×10 -7 , 5.9×10 -7 , and 2.6×10 -6 ft 3 hr -1 gal -1 , respectively. The Limit of Detection and Limit of Quantification for the HGR measurements were 2.6×10-8 ft3 hr-1 gal-1 and 5.1×10 -8 ft 3 hr -1 gal -1 , respectively. HGR was measured in a high boiling point (HBP) simulant with major known SRS Tank Farm organic compounds. Testing used fresh organics and did not simulate the long-term exposure of these organics already experienced in the waste. Hence, testing is expected to provide high bias to the hydrogen generation rates. The final HGR measurements at 75, 100, and 140 °C were 6.2×10 -7 , 3.0×10 -6 , and 8.1×10 -5 ft 3 hr -1 gal -1 , respectively. Decreases in the HGR were noted at isothermal conditions with steady gas tracer concentration, which clouds development of a temperature dependent correlation for generation rate. While the simulant contains high concentrations of major SRS organics tested at high temperatures found in the 3H Evaporator, this limited data set may not fully bound values under all possible operating conditions. The HGR measurement apparatus was successfully employed to replicate hydrogen production rate experiments. Simulant without deliberately added organic carbon was found to contain approximately 200 mg/L of organic carbon from tramp organics in the reagents used to prepare the solution. The tests using such simulant generated hydrogen at rates consistent with, but somewhat lower than, those predicted at equivalent carbon and aluminum loadings by methods used at the Hanford Site. Prior to identifying the organic contaminants, potential sources of hydrogen in simulants without deliberately added organic carbon were examined. Testing determined that this hydrogen is not due to caustic reaction of organic materials of construction within the reaction vessel, electrolysis due to current leak, anode/cathode reactions, corrosion, or the method of heating employed in the experiment. During the HBP and Tank 38 testing, a peak was identified at a retention time greater than the retention time of the inert tracer of krypton. This peak has been identified as methane, and the data have been reexamined to obtain methane concentrations. For the HBP testing, methane concentration was constant at the highest temperature studied (140 °C) and higher than hydrogen. At the lower temperatures (100 and 75 °C), methane and hydrogen tracked closely, with methane concentration generally lower than hydrogen. For Tank 38, methane concentration was approximately 1/3 of the hydrogen concentration during testing (100 and 110 °C). At lower temperatures for Tank 38 testing (90, 80, and 27 °C), methane fell below the quantification range.

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DOE OSTI2020

Characterization of Tank 11H samples from tank closure cesium removal (TCCR) Batch 3 - intermediate and final samples

Savannah River Remediation (SRR) is currently operating the Tank Closure Cesium Removal (TCCR) process to remove 137 Cs from tank waste supernate using an ion exchange process. The TCCR unit processes dissolved salt from Tank 10H through a series of ion exchange columns containing crystalline silicotitanate (CST) and the effluent is then discharged to Tank 11H. Four interim samples pulled from Tank 11H during and just after the completion of processing of Batch 3 through the TCCR process have been analyzed for 137 Cs activity and density. The 137 Cs activity was found to decrease with each subsequent sample, which is consistent with the addition of decontaminated solution to Tank 11H. When compared to the expected composition from mixing the Tank 10H Batch 3 feed with the material already present in Tank 11H, the bulk chemical composition was as expected. A corrosion control sample collected from Tank 11H in June 2020 showed changes in the chemical composition and 137 Cs activity when compared to the composition measured at the end of Batch 2 processing. As there were no additions made to the tank during this period, these changes were attributed to leaching of the solids present in Tank 11H. Revision 1 of this report contains additional analyses of the fourth interim sample including inductively coupled plasma – mass spectrometry (ICP-MS) results and activities of other radionuclides. These results are consistent with the previously reported results, showing leaching from the Tank 11H heel. Notably, the 90 Sr activity was about 2.7x higher than calculated from a mixture of the original Tank 11H supernate and the Tank 10H supernate as treated by TCCR, suggesting additional leaching of Sr from the Tank 11H heel solids. In addition, Revision 1 includes analysis results for the Batch 3 Post-Production surface and variable depth samples collected in October of 2020. The Post-Production surface sample was similar in composition to the fourth interim surface sample, except for the increased Cs-137 and oxalate concentrations that were presumably associated with leaching of the Tank 11H heel solids.

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DOE OSTI2024

Preliminary Analysis of Advanced Reactor Spent Nuclear Fuel Storage, Transportation, and Disposal

Due to increased interest in advanced reactor deployment and their associated potential new fuel cycles, the U.S. Department of Energy (DOE) Spent Fuel and Waste Science and Technology (SFWST) program has begun to evaluate the possible implications of long term management and final disposition of the spent nuclear fuel (SNF) generated. Safely managing and dispositioning this SNF, along with any other associated radioactive waste, is the primary focus of this initial preliminary assessment. This paper summarizes efforts to evaluate the characteristics and packaging options for three types of advanced reactor SNF forms: (1) tristructural isotropic (TRISO), (2) metallic, and (3) irradiated fuel salt presented in the report titled “Storage, Transportation, and Disposal of Advanced Reactor Spent Nuclear Fuel and High-Level Waste”. TRISO and metallic SNF and their associated waste streams were emphasized because of the near-term anticipated demonstrations of X-energy’s Xe-100 and TerraPower and GE Hitachi’s Natrium advanced reactors. Preliminary information on spent fuel salts discharged from molten-salt reactors (MSRs) was also examined to provide a baseline for future efforts. All calculations and assumptions were based on publicly available information. This paper identifies several different reactors that produce either TRISO or metallic SNF as well as a few of the reactor and fuel characteristics used for the assessments. Based on these characteristics, calculations were performed to determine the applicability of packaging SNF into existing canister designs. The evaluations included geometric (e.g., dimension, volume) and mass/weight considerations, known operational approaches and loading procedures, physical and chemical considerations/conditions for storage environments, and as-loaded radiation, thermal, and criticality analyses to identify constraints on storage, transportation, and disposal. Gaps in publicly available data pertaining to reactor operation and/or fuel composition provide increased uncertainty in some evaluations. Additionally, uncertainty in packaging and SNF management operations provide additional uncertainty. However, preliminary conclusions can still be assessed through this work and are presented in this paper.

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